Presentation is loading. Please wait.

Presentation is loading. Please wait.

Y. Kamada JAEA TCM-8 meeting, 2010, April 14 ENEA Frascati, Italy 1 JT-60SA Plasma Regimes and Research Plan.

Similar presentations


Presentation on theme: "Y. Kamada JAEA TCM-8 meeting, 2010, April 14 ENEA Frascati, Italy 1 JT-60SA Plasma Regimes and Research Plan."— Presentation transcript:

1 Y. Kamada JAEA TCM-8 meeting, 2010, April 14 ENEA Frascati, Italy 1 JT-60SA Plasma Regimes and Research Plan

2 The JT-60SA Project The JT-60SA project is conducted under the BA Satellite Tokamak Programme by Europe and Japan, and the Japanese National Programme. JT-60SA and ITER should be operated as a ‘set’, in order to realize the Fusion Energy for both * science and technology * scientists The project mission is to contribute to early realization of fusion energy by supporting exploitation of ITER and by complementing ITER with resolving key physics and engineering issues for DEMO reactors. 2

3 JT-60SA Plasma Regimes JT-60SA is a fully superconducting tokamak capable of confining break-even equivalent class high-temperature deuterium plasmas (Ip-max=5.5 MA) lasting for a duration (typically 100s) longer than the timescales characterizing the key plasma processes, such as current diffusion and particle recycling. JT-60SA should pursue full non-inductive steady-state operations with high  N (> no-wall ideal MHD stability limits). 3 Sustainment Time (s) JT-60SA Target DEMO reactors ITER NN Existing Tokamaks JT-60U Steady-state Inductive

4 EU & JA Share Procurements Design of the key components has been almost completed. The JT-60SA project has been entered its manufacturing stage. Laser scattering Cryostat TF coils&Testing Assembly NBI In-vessel Components Remote Handling Vacuum Vessel Diagnostics CS, EF coils Disassembly Rad. safety Cont/data Magnet Interface Power Supplies Cryogenic System ECRF Compressor Building Water Cooling System Naka site 16m Base: 260t Body: 350t 150t 700t 4

5 5 Project Schedule

6 JT-60SA: Highly Shaped Large Superconducting Tokamak A wide range of plasma equilibrium covering a high plasma shaping factor of S =q 95 I p /(aB t ) ~7 and a low aspect ratio of A~2.5 with a sufficient inductive plasma current flattop and additional heating up to 41 MW during 100 s. The plasma size is ~ 0.5 x ITER = between ITER and other superconducting tokamaks. An integrated knowledge of superconducting tokamaks SST-1, EAST, KSTAR, TORE-SUPRA, JT-60SA and ITER will establish a reliable nuclear fusion science and technology towards DEMO. 6

7 7 Typical Plasma Parameters Ip=5.5MA, Double Null Ip=4.6MA ITER- shape

8 8 Ip=5.5MA Discharge Example

9 9 High  N, high bootstrap Steady-state operation

10 10 High  N steady-state operation space

11 Research needs for ITER & DEMO 11

12 12

13 Fully Superconducting Large Tokamak Highly Shaped Plasma Configuration Strong Heating and Current Drive Power with Variety & Long Pulse Large Capability of Stability Control Large Capability of Divertor Power Handling and Particle Control Variety of High Resolution Diagnostics JT-60SA device has been designed in order to satisfy the central research needs for ITER and DEMO JT-60SA device 13

14 41MWx100s High Power Heating with Variety Positive-ion-source NB 85keV 12units x 2MW=24MW COx2u, 4MW CTRx2u, 4MW Perpx8u, 16MW Negative-ion-source NB 500keV, 10MW Off-axis for NBCD variety of heating/current-drive/ momentum-input combinations ECRF: 110GHz, 7MW x 100s 9 Gyrotrons, 4 Launchers with movable mirror >5kHz modulation NB: 34MWx100s 14 ECRF NB

15 15 In-vessel components for stability control  RFX

16 16 Divertor Structure for heat & particle control

17 Research phases and status of key components 17 JT-60SA operation starts earlier than ITER’s hydrogen operation by ~5 years. The tight schedule of ITER up to DT1 requires sufficient explorations of the key physics and operational techniques in satellite devices. => Experiences and achievements in JT-60SA are indispensable for smooth and reliable progress of ITER.

18 Divertor & Wall Material Research for DEMO Extended Research Phase: Installation of the metallic divertor targets and first wall together with an advanced shape divertor will be conducted based on progress of the research in the world tokamaks including ITER. Integrated Research Phase: The material of the divertor target and the first wall is now considered to be carbon before achievement of the JT-60SA’s main mission of the high-  steady-state. However, possibility of replacement to metallic materials will be discussed based on the results in JET, ASDEX-U, FTU. Replaceable Divertor Cassette 18

19 19 JT-60SA Research Plan

20 ITER & DEMO-relevant non-dimensional regime JT-60SA allows explorations in the ITER- and DEMO-relevant plasma regimes in terms of non-dimensional plasma parameters at high plasma densities in the range of 1×10 20 /m 3. 20

21 ITER & DEMO-relevant heating condition / scan ECH (110GHz, 7MW) N-NB (500keV, 10MW) => Electron Heating dominant Low Particle input Low Torque input P-NB (85keV, 24 MW) => Ion Heating dominant Perp-NB & balanced CO/CTR-NB => low torque input ( torque input scan) JT-60SA allows dominant electron heating, scan of power ratio to electron high power heating with low central fueling high power heating with low external torque input ( incl. scan of rotation) 21

22 Study on highly self-regulating plasmas for DEMO High beta & high bootstrap fraction => strong linkages among j(r), p(r), Vt(r) + Global linkage / Global structure including core & pedestal + Linkage among transport coefficients & roles of MHD activities => JT-60SA allows understanding & control of this plasma system at ITER- & DEMO- relevant non-dimensional parameters (  *, *,  N,  p, q 95 …) 22 JT-60SA plasma actuator system allows separated controls for heating, current drive, rotation drive & fueling.

23 23 RWM control coils Stabilizing plate JT-60SA allows exploitations of high beta regimes with the high shape factor S up to 7, the stabilizing shell, the RWM control coils, the error field correction coils, and the high power heating & CD & momentum-input. Demonstration & Study of High Beta (>non-wall limit) for DEMO For DEMO, minimum rotation for RWM stabilization has to be studied => w/o control coils. Identification of the disruption limits at high  N.  N critical = 4.32  N no-wall = 3.12  N ideal-wall =4.40

24 JT-60SA supports ITER’s main mission & commissioning with high Ip, high power, high density plasmas * H-mode operations towards Q=10 L-H transition Pedestal Structure H-mode confinement ( incl. compatibility with radiative divertor, RMP, etc.) Disruption behavior, and disruption prediction & mitigation using the same techniques planed in ITER. Operation scenario optimization with superconducting PF coils. Divertor heat load mitigation ( incl. ELMs) and particle controllability JT-60SA has sufficient power for L-H transition & H-mode confinement studies at Ip=5.5MA & ne=10 20 m -3. JT-60SA provides data & techniques for 24 with 10MW high energy (500keV) N-NB; NB Current Drive studies (incl. off-axis NBCD), AE mode stability & effects on fast-ion transport, Interactions between high energy ions and MHD instabilities

25 ELM mitigation for ITER and DEMO JT-60SA’s high Ip high power H-mode plasmas allow type I ELM studies at sufficiently low edge collisionality. JT-60SA ELM energy loss fraction Error field correction / generation coils are used for RMP 2) JT-60SA’s high triangularity plasmas allow small ELM regimes ( i.e. grassy ELM) for DEMO. (DEMO-equivalent shape ) => ELM mitigation without RMP. 1)ELM mitigation by RMP & pellet inj. for ITER. collisionality 25 JT-60SA

26 JT-60SA allows exploitations of NB Current Drive studies (incl. off-axis NBCD), AE mode stability & effects on fast-ion transport, Interactions between high energy ions and MHD instabilities with 10MW high energy (500keV) N-NB. High Energy Particle Studies for ITER & DEMO 26

27 27 The peak heat flux can be suppressed within the mono block capability (15 MW/m 2 ) by gas puffing for 41 MW injection. (n e,ave ~1x10 20 m -3 at f GW =0.8). Divertor Power Handling for ITER & DEMO Radiation map == SONIC code simulation == 10.4 MW/m 2 Heat flux density on the LFS target The ITER-like W-shaped divertor with a V-corner enhances divertor radiation. CFC monoblock divertor target allows 15MW/m 2. Test at JEBIS at 15MW/m2 for 12 full-size mock- ups of monoblock target with 10 CFC blacks. About half of mock-ups satisfied the requirements At lower n e compatible with lower I p plasmas, q peak = 8.6 MW/m 2 is obtained with impurity seeding. (30x30x30mm) Qualified targets survived 2000 heat cycles

28  Divertor condition can be controlled from attached to detached conditions with constant main plasma density. Divertor pumping with cryopumps allows Pumping speed of m 3 /s by 8 steps. Fuel & Impurity Particle Control for ITER & DEMO Compatibility of the radiative divertor with impurity seeding and sufficiently high fuel purity in the core plasma should be demonstrated. The key point is to clarify whether a wide range of the divertor plasma controllability can be realized independently of the main plasma operation condition. JT-60SA demonstrates particle controls under saturated wall condition by utilizing variety of the fuelling and pumping systems (gas-puffing from man and divertor, pellet injection, divrtor pumping).

29 High Integrated Performance for DEMO ‘Simultaneous & steady-state sustainment of the key performances required for DEMO’ has never been achieved => the goal of JT-60SA. Example of JT-60SA at Ip=2.3MA. 29

30 Integrated Control Scenario Development Understanding & Control of the highly self-regulating combined plasma system for DEMO High-beta, high-bootstrap fraction plasma => a highly self regulating non-linear system governed by strong linkages among j(r), p(r) and Vt(r) in core & pedestal. Strong spatial linkage : Core – Pedestal – SOL – Divertor plasmas Study controllability & Plasma response Determine the minimum suitable set of actuators & logic. + control margin against operation boundaries (In particular disruption limits ) 30

31 Summary The project mission of JT-60SA is to contribute to early realization of fusion energy by supporting exploitation of ITER and by complementing ITER for DEMO. JT-60SA device has been designed in order to satisfy all of the central research needs for ITER and DEMO, in particular, ‘Simultaneous & steady-state sustainment of the key performances required for DEMO’ & ‘Integrated Control Scenario Development ‘. 31

32 JT-60SA is indispensable for ITER & DEMO JT-60SA operation starts earlier than ITER’s hydrogen operation by ~5 years. The tight schedule of ITER up to DT1 requires sufficient explorations of the key physics and operational techniques in satellite devices. => Experiences and achievements in JT-60SA are indispensable for smooth and reliable progress of ITER. For DEMO, an integration of achievements in JT-60SA high-  steady-state plasmas and ITER burning plasmas is required to make DEMO designs more realistic and attractive. For early realization of DEMO, such integrated exploitation of JT-60SA and ITER is necessary. 32

33 33

34

35

36


Download ppt "Y. Kamada JAEA TCM-8 meeting, 2010, April 14 ENEA Frascati, Italy 1 JT-60SA Plasma Regimes and Research Plan."

Similar presentations


Ads by Google